Merge pull request #62 from fusion-energy/allowing_tallies_to_be_comb…

…ined

allowing multiple tallies to be added and plotted

.github/workflows/ci_with_install.yml

@@ -44,6 +44,8 @@ jobs:
       - name: Run examples
         run: |
           cd examples
+          python plot_minimal_2d_example.py
           python plot_minimal_example.py
-          python plot_with_custom_color_map.py
+          python plot_two_tallies_combined.py
           python plot_sweep_through_slice_indexes.py
+          python plot_with_custom_color_map.py

README.md

@@ -21,11 +21,10 @@ Features
 
 :arrow_right_hook: Customisable by passing keywords to underlying matplotlib functions colorbar, contour and imshow
 
-:arrow_right_hook: supports further customisable by supporting ```matplotlib.rc```
-```python
-import matplotlib 
-matplotlib.rc('font', **{'family' : 'normal', 'size'   : 22})
-```
+:arrow_right_hook: supports further customisations throught ```matplotlib.rc()```
+
+:heavy_plus_sign: Add tally results together to get combined plot.
+
 |<img src="https://user-images.githubusercontent.com/8583900/265032335-27463ee9-8960-4f5e-a662-dab0b6cd9fc5.png" alt="drawing" width="400"/>|<img src="https://user-images.githubusercontent.com/8583900/265065370-734c66ab-b20e-40c8-b72b-88203ea4347b.gif" alt="drawing" width="400"/>|
 
 # Local install

examples/plot_sweep_through_slice_indexes.py

@@ -7,6 +7,10 @@
 from matplotlib.colors import LogNorm
 from openmc_regular_mesh_plotter import plot_mesh_tally
 from matplotlib import cm
+import matplotlib
+
+# sets the font for the axis
+matplotlib.rc("font", **{"family": "normal", "size": 22})
 
 # MATERIALS
 

examples/plot_two_tallies_combined.py

@@ -0,0 +1,122 @@
+import openmc
+from matplotlib.colors import LogNorm
+from openmc_regular_mesh_plotter import plot_mesh_tally
+
+
+# MATERIALS
+mat_1 = openmc.Material()
+mat_1.add_element("Li", 1)
+mat_1.set_density("g/cm3", 0.1)
+my_materials = openmc.Materials([mat_1])
+
+# GEOMETRY
+# surfaces
+inner_surface = openmc.Sphere(r=200)
+outer_surface = openmc.Sphere(r=400, boundary_type="vacuum")
+# regions
+inner_region = -inner_surface
+outer_region = -outer_surface & +inner_surface
+# cells
+inner_cell = openmc.Cell(region=inner_region)
+outer_cell = openmc.Cell(region=outer_region)
+outer_cell.fill = mat_1
+my_geometry = openmc.Geometry([inner_cell, outer_cell])
+
+# SIMULATION SETTINGS
+my_settings = openmc.Settings()
+my_settings.batches = 10
+my_settings.inactive = 0
+my_settings.particles = 50000
+my_settings.run_mode = "fixed source"
+# my_settings.photon_transport = True # could be enabled but we have a photon source instead which converges a photon head deposition quicker
+
+# Create a neutron and photon source
+try:
+    source_n = openmc.IndependentSource()
+    source_p = openmc.IndependentSource()
+except:
+    # work with older versions of openmc
+    source_n = openmc.Source()
+    source_p = openmc.Source()
+
+source_n.space = openmc.stats.Point((200, 0, 0))
+source_n.angle = openmc.stats.Isotropic()
+source_n.energy = openmc.stats.Discrete([0.1e6], [1])
+source_n.strength = 1
+source_n.particle = "neutron"
+
+source_p.space = openmc.stats.Point((-200, 0, 0))
+source_p.angle = openmc.stats.Isotropic()
+source_p.energy = openmc.stats.Discrete([10e6], [1])
+source_p.strength = 10
+source_p.particle = "photon"
+
+my_settings.source = [source_n, source_p]
+
+# Tallies
+mesh = openmc.RegularMesh().from_domain(
+    my_geometry,  # the corners of the mesh are being set automatically to surround the geometry
+    dimension=[40, 40, 40],
+)
+
+mesh_filter = openmc.MeshFilter(mesh)
+neutron_filter = openmc.ParticleFilter("neutron")
+photon_filter = openmc.ParticleFilter("photon")
+
+mesh_tally_1 = openmc.Tally(name="mesh_tally_neutron")
+mesh_tally_1.filters = [mesh_filter, neutron_filter]
+mesh_tally_1.scores = ["heating"]
+
+mesh_tally_2 = openmc.Tally(name="mesh_tally_photon")
+mesh_tally_2.filters = [mesh_filter, photon_filter]
+mesh_tally_2.scores = ["heating"]
+
+my_tallies = openmc.Tallies([mesh_tally_1, mesh_tally_2])
+
+
+model = openmc.model.Model(my_geometry, my_materials, my_settings, my_tallies)
+sp_filename = model.run()
+
+# post process simulation result
+statepoint = openmc.StatePoint(sp_filename)
+
+# extracts the mesh tally by name
+my_mesh_tally_photon = statepoint.get_tally(name="mesh_tally_photon")
+my_mesh_tally_neutron = statepoint.get_tally(name="mesh_tally_neutron")
+
+# default tally units for heating are in eV per source neutron
+# for this example plot we want Mega Joules per second per cm3 or Mjcm^-3s^-1
+neutrons_per_second = 1e21
+eV_to_joules = 1.60218e-19
+joules_to_mega_joules = 1e-6
+scaling_factor = neutrons_per_second * eV_to_joules * joules_to_mega_joules
+# note that volume_normalization is enabled so this will also change the units to divide by the volume of each mesh voxel
+# alternatively you could set volume_normalization to false and divide by the mesh.volume[0][0][0] in the scaling factor
+# in a regular mesh all the voxels have the same volume so the [0][0][0] just picks the first volume
+
+plot = plot_mesh_tally(
+    tally=[my_mesh_tally_neutron],
+    colorbar=True,
+    # norm=LogNorm()
+)
+plot.title.set_text("neutron heating")
+plot.figure.savefig("neutron_regular_mesh_plotter.png")
+print("written file neutron_regular_mesh_plotter.png")
+
+plot = plot_mesh_tally(
+    tally=[my_mesh_tally_photon],
+    colorbar=True,
+    # norm=LogNorm()
+)
+plot.title.set_text("photon heating")
+plot.figure.savefig("photon_regular_mesh_plotter.png")
+print("written file photon_regular_mesh_plotter.png")
+
+plot = plot_mesh_tally(
+    tally=[my_mesh_tally_photon, my_mesh_tally_neutron],
+    colorbar=True,
+    # norm=LogNorm()
+)
+plot.title.set_text("photon and neutron heating")
+plot.figure.savefig("photon_and_neutron_regular_mesh_plotter.png")
+print("written file photon_and_neutron_regular_mesh_plotter.png")